Processing ultra high temperature zirconium carbide microencapsulated nuclear fuel

ABSTRACT

The known fully ceramic microencapsulated fuel (FCM) entrains fission products within a primary encapsulation that is the consolidated within a secondary ultra-high-temperature-ceramic of Silicon Carbide (SiC). In this way the potential for fission product release to the environment is significantly limited. In order to extend the performance of this fuel to higher temperature and more aggressive coolant environments, such as the hot-hydrogen of proposed nuclear rockets, a zirconium carbide matrix version of the FCM fuel has been invented. In addition to the novel nature to this very high temperature fuel, the ability to form these fragile TRISO microencapsulations within fully dense ZrC represent a significant achievement.

CROSS-REFERENCES TO RELATED APPLICATIONS

-   -   This application claims priority to U.S. Provisional Patent        Application No. 62/774,303, filed Dec. 2, 2018, titled        PROCESSING ULTRA HIGH TEMPERATURE ZIRCONIUM CARBIDE        MICROENCAPSULATED NUCLEAR FUEL, the entire disclosure of which        is incorporated by reference herein.    -   The entire disclosures of U.S. Pat. No. 9,299,464 B2 “Fully        Ceramic Nuclear Fuel and Related Methods” and U. S. Patent        2017/0025192 A1 “Method for Fabrication of Fully Ceramic        Microencapsulated Nuclear Fuel” are incorporated by reference        herein.

BACKGROUND OF THE INVENTION 1. Field of the Invention

This invention relates to a nuclear fuel. More specifically thisinvention describes a new fuel form and the method for fabrication ofthis inert matrix fuel whereby a fragile fuel microencapsulation isconsolidated within an ultra-high-temperature-ceramic (UHTC) (zirconiumcarbide) that serves as a secondary and very high-temperature barrier tofission product release.

2. Description of the Related Art

There are many known types of nuclear fuel for both research, powerproducing nuclear reactors, and reactor for space exploration. The mostcommon example of nuclear fuel is the ceramic uranium oxide pellet thatis contained within a thin metallic cladding. That cladding bothprovides a rigid structure to hold the fuel and serves as the barrier tofission product release to the coolant stream. A second example ofnuclear fuel is an inert matrix fuel (IMF) in which a fissile materialsuch as (or containing) U-235 is dispersed in an inert host matrix. Thatinert matrix may be a refractory ceramic and is intended to provide arigid host for the fuel as well as provide a measure of fission productretention. Yet a third example is a microencapsulated fuel, in oneexample known as the fully ceramic microencapsulated FCM™ fuel. Thisfuel type, similar to that of the IMF, has a distinct non-fuel matrixsurrounding a plurality of fueled particles otherwise known asmicroencapsulations. In contrast with the IMF the microencapsulatedfuels utilize an engineered fuel microencapsulation, such as thetri-isotropic (TRISO) or bi-isotropic (BISO) fuel forms which layerspyrolitic graphite and SiC (for the TRISO) around the fissile fuelkernels thus providing barriers to fission product retention within thefuel. The historic and most common porous host matrix is graphite. Sucha fuel was developed as early nuclear thermal propulsion rockets withgood success. This FCM fuel has the attribute of having both a primaryfission product barrier (the TRISO fuel particle) and a secondarybarrier being the SiC matrix. This combination thus provides two ruggedbarriers to fission product release.

The SiC matrix of the known FCM™ fuel is fabricated through a TransientEutectic-Phase (TEP) process whereby rare earth oxides are utilized toreduce the sintering temperature and pressure required to achieve fulldensity for the SiC matrix. As taught in U.S. Pat. No. 9,299,464 B2 and2017/0025192 A1 this process is essential in allowing a process windowfor which the fragile TRISO particles will not be crushed and therebyrendered ineffective in a fuel application. However, while the standardFCM product is considered robust for application temperatures up to1850° C., above that temperature unacceptable SiC matrix degradationoccurs due to instability of TEP SiC. Moreover, for application insystems such as nuclear thermal rocket engines the reaction between SiCand hot hydrogen is unacceptable. In order to move into a higherperformance regime, a higher temperature ultra-high temperature ceramic(UHTC) that replaces the SiC matrix of the microencapsulated fuel is putforward. Typical UHTC's and their maximum application temperatures areprovided in the table below. Of those materials listed zirconium carbidehas a number of attractive features as an ultra-high temperature fuelmatrix, though has historically been very difficult to process,requiring temperature and pressure well in excess of that which wouldcrush modern TRISO fuel microencapsulation. The present inventionprovides a process to fabricate a zirconium carbide matrixmicroencapsulated fuel at conditions favorable to the use of TRISO fuel.

Melt or Decomp. Crystal Density Temperature Material Formula structure(g/cm³) (° C.) Niobium nitride NbN Cubic 8.470 2573 Tantalum nitride TaNCubic 14.30 2700 Vanadium carbide VC Cubic 5.77 2810 Silicon carbide SiCCubic 3.21 2820 Zirconium nitride ZrN FCC 7.29 2950 Titanium nitride TiNFCC 5.39 2950 Tantalum boride TaB₂ HCP 12.54 3040 Titanium carbide TiCCubic 4.94 3100 Titanium boride TiB₂ HCP 4.52 3225 Zirconium boride ZrB₂HCP 6.10 3245 Hafnium boride HfB₂ HCP 11.19 3380 Hafnium nitride HfN FCC13.9 3385 Zirconium carbide ZrC FCC 6.56 3400 Niobium carbide NbC Cubic7.820 3490 Tantalum carbide TaC Cubic 14.50 3768 Hafnium carbide HfC FCC12.76 3958

BRIEF SUMMARY OF THE PRESENT INVENTION

The present invention provides the concept for azirconium-carbide-matrix, ultra-high-temperature ceramic matrix fueldesigned for advanced nuclear application. As envisioned, in comparisonwith standard FCM fuel which would achieve a nominal specific impulse inthe range of 500-600 s, the use of ZrC-based or pure ZrC matrix UHTC FCMfuel would achieve in the range of 700-850 s. Application of this fuelincludes nuclear rocket engines and systems requiring fuel of limitedfission product release to operate at temperatures in excess of 2500° C.Use of ZrC matrix UHTC FCM could incorporate TRISO of standard or morerugged SiC micropressure vessels for short durations in the temperaturerange of 2200-2400° C., 2 hr time periods. Advanced microencapsulationwhereby ZrC shells replace the SiC microencapsulation of the TRISO canbe considered for longer life or higher temperature application. TheTRISO containing ZrC is known to the literature (i.e. TRIZO), displayingsimilar crush strength to standard modern TRISO.

The critical step towards achieving fabrication of the ZrC matrix UHTCFCM is the ability to consolidate a near full density matrix of ZrCwhile not compromising the function of the entrained second phasefissile fuel: to not crush, deform, or substantially react layers. Inthe case of a SiC (or ZrC) TRISO microencapsulation this meansconsolidation without rupture of a significant (<10 ppm) number of theSiC (or ZrC) protective shell layers. This is accomplished throughsuppression of the normal pressures and temperatures required forsintering of zirconium carbide. In addition to a low failure fraction ofTRISO particles a measure of success for the matrix is to achieve nearfull density without interconnected porosity. As described, two methodshave achieved acceptable levels of success in producing ZrC matrix UHTCFCM: A) Transient Eutectic-Phase Processing, B) Hydrogen AidedSintering.

BRIEF DESCRIPTION OF THE DRAWINGS

FIG. 1 is a schematic of the (known) FCM fuel geometry. The TRISO fuelform (right-most) is embedded in a matrix of fully dense Ultra HighTemperature Ceramic. Form factor is arbitrary.

REFERENCE NUMERALS IN THE DRAWINGS

-   1 Microencapsulated fuel    -   1A Fissile fuel kernel    -   1B Outer pyrolitic carbon layer of microencapsulated fuel    -   1C SiC or alternate UHTC layer of microencapsulated fuel    -   1D Inner pyrolitic carbon layer of microencapsulated fuel    -   1E Buffer graphitic layer of microencapsulated fuel-   2 Ceramic fuel sleeve-   3 FCM mixture to be cold pressed    -   3A FCM constituent mixture: Zr and C powder, microencapsulated        fuel, silica, aluminum oxide, and/or neutron poison rare earth        oxides.    -   3B FCM constituent mixture: Zr and C powder, yttrium oxide,        aluminum oxide, and neutron poison rare earth oxide.

DETAILED DESCRIPTION OF THE INVENTION

FIG. 1 provides a schematic similar to the known fully ceramicmicroencapsulated (FCM.) Whereby the known FCM utilizes SiC powder as amajor matrix constituent, FIG. 1 portrays a fuel that is comprised ofthe fissile-fuel containing microencapsulation, in this case depicted asa Tri-Isotropic (TRISO) particle (item 1.) The new ZrC-matrix UHTC FCMfuel is demonstrated with acceptable matrix density and encapsulationintegrity, post-processing, as achieved by the following means:

Transient Eutectic-Phase Approach: Utilizing a combination of kineticball milling and high-shear milling to combine a certain volume fractionof ZrC powder and small percentage of SiC and oxide (such as Al₂O₃ andY₂O₃) powders. This combination of ZrC, SiC and oxide powders form thedense matrix of this UHTC microencapsulated fuel. The compact of FIG. 1,forming the UHTC-FCM are fabricated at pressures not exceeding 20 MPaand temperatures not exceeding 2000° C. to attain a continuous, lowporosity, ZrC matrix surrounding TRISO particles which remain unbrokenand intimately bonded with the matrix following processing. The amountof oxide eutectic aids in the starting powder mix for processing theUHTC FCM fuel is up to 3 weight percent. The amount of SiC in thestarting powder mix is up to 30 weight percent.

Specifically, the transient eutectic phase (TEP) SiC mixture iscomprised using 94% SiC, 3.9% Al2O3, 2.1% Y2O3 by mass. The density wascritically sensitive to both amount and ratio of rare earth additives.The SiC powder used were either 35 nm or 85 nm nanophase powder producedby chemical vapor deposition process. The TEP SiC mixture was mixed withmilling media, dried, deagglomerated and re-dispersed in atmosphereprior to use. The feedstock material consists of powders sourced fromcommercial vendors and processed under typical methodologies found inceramic powder forming and sintering. The TEP SiC mixture was added toZrC in ratios of ZrC with 10 wt % TEP SiC mixture and re-mixed aspreviously described. The feedstock powder is mixed with a proprietaryset of dispersants, binders, flow plasticizers and release agents, whichassist in rheological properties needed for forming operations.

Sintering was conducted inside graphite tooling, configured for ˜10 mmdiameter. Green bodies are formed in-die by loading the powder mixturedirectly into prepared graphite tooling. A cylindrical graphite die witha cylindrical cavity was lined with graphite foil. This conducts heatinto the pellet and provides a release interface for the consolidatedpellet. The pellet is formed by pouring the ZrC-based powder, followedby compaction by spark plasma sintering (SPS) for 10 minutes at 20 MPaat room temperature during the vacuum cycle. Before sintering thepressure was reduced to 10 MPa. The applied pressure was limited inorder to establish processing conditions compatible with TRISOparticles, which fail at low pressures at room temperature. Sinteringtemperatures of 1875° C. for 10 minutes. After sintering, the ZrC-10%TEP SiC mixture made into solid pellets. The pellets achieved 93%theoretical density.

-   -   Hydrogen Aided Reaction Sintering: Utilizing a        non-stoichiometric mixture of ZrC powder, ZrH and carbon powder        the UHTC FCM ZrC matrix takes advantage of enhanced diffusion in        a sub-stoichiometric ZrC and the decomposition of ZrH at        approximately 900° C. In the absence of hydride decomposition        the temperature and pressures were in excess of 2000° C. and 60        MPa with <95% theoretical density. With the addition of percent        levels of ZrH the compact achieves near full density at        temperatures under 1800° C. Powder handling and direct current        sintering is carried out in a similar fashion to the Transient        Eutectic-Phase Approach. Pressures not exceeding 20 MPa and        temperatures in the 1650-1800° C. range produce a dense matrix        and rupture-free TRISO microencapsulations. ZrH additions up to        10 weight percent by mass are demonstrated effective with free        carbon in the range of 0.1 to 4% by mass.

Having described the invention, I claim:
 1. A nuclear fuel productcomprising: fragile fuel particles that include a fuel kernel and one ormore layers encapsulated within a zirconium carbide (ZrC) matrix,wherein the zirconium carbide matrix includes near fully dense ZrC thatis sub-stoichiometric ZrC between 93% theoretical density and less than100% theoretical density.
 2. The nuclear fuel product of claim 1,wherein the fragile fuel particles include Tri Structural Isotropic fuel(TRISO) fuel particles or a variation of TRISO fuel particles utilizingzirconium carbide (TRIZO).
 3. The nuclear fuel product of claim 1,wherein the fragile fuel particles include Bi Structural Isotropic fuel(BISO) fuel particles.
 4. A method of fabricating the nuclear fuelproduct of claim 1, comprising: utilizing a non-stoichiometric reactionof ZrC, zirconium hydride (ZrH), and free carbon (C), wherein the ZrC is0-10 mass percent and the free carbon is 0-4 mass percent.
 5. A nuclearthermal propulsion system, comprising: the nuclear fuel product ofclaim
 1. 6. A nuclear reactor, comprising: the nuclear fuel product ofclaim
 1. 7. A method of fabricating the nuclear fuel product of claim 1,comprising: creating a transient eutectic phase (TEP) silicon carbide(SiC) mixture that includes an SiC powder and an oxide power of at leastone sintering aid to suppress a processing temperature for sintering ofZrC.
 8. The method of fabricating the nuclear fuel product of claim 7,further comprising: prior to creating the TEP SiC mixture, implementinga chemical vapor deposition process to form the SiC powder as a 35nanometer (nm) to 85 nm nanophase powder.
 9. The method of fabricatingthe nuclear fuel product of claim 7, wherein the TEP SiC mixtureincludes approximately 94% SiC, 3.9% aluminum oxide (Al₂O₃), and 2.1%yttrium oxide (Y₂O₃) by mass.
 10. The method of fabricating the nuclearfuel product of claim 7, further comprising: mixing a ZrC powder withthe TEP SiC mixture to form a ZrC-TEP SiC mixture.
 11. The method offabricating the nuclear fuel product of claim 10, wherein the mixing theZrC powder with the TEP SiC mixture to form the ZrC-TEP SiC mixtureincludes: adding the TEP SiC mixture to the ZrC powder; and kineticmilling and high-shear milling to combine the ZrC powder with the TEPSiC mixture.
 12. The method of fabricating the nuclear fuel product ofclaim 10, wherein the mixing the ZrC powder with the TEP SiC mixture isin a ratio of 10 wt % TEP SiC mixture.
 13. The method of fabricating thenuclear fuel product of claim 10, further comprising: adding the fragilefuel particles to the ZrC-TEP SiC mixture; loading the ZrC-TEP SiCmixture with the added fragile fuel particles into a die; and compactingthe ZrC-TEP SiC mixture and the fragile fuel particles by spark plasmasintering.
 14. The method of fabricating the nuclear fuel product ofclaim 13, wherein the spark plasma sintering is at a processing pressureof less than or equal to 20 megapascals (MPa) and the suppressedprocessing temperature is less than or equal to 2,000 degrees Celsius.15. The method of fabricating the nuclear fuel product of claim 13,wherein after the spark plasma sintering: the zirconium carbide matrixsurrounds the fragile fuel particles; and the fragile fuel particles areunbroken and are intimately bonded with the zirconium carbide matrix.16. A method of fabricating the nuclear fuel product of claim 1,comprising: creating a non-stoichiometric mixture that includes a ZrCpowder, a carbon (C) powder, and zirconium hydride (ZrH) to suppress aprocessing temperature for sintering of ZrC.
 17. The method offabricating the nuclear fuel product of claim 16, further comprising:adding the fragile fuel particles to the non-stoichiometric mixture; andsintering the non-stoichiometric mixture and the fragile fuel particles.18. The method of fabricating the nuclear fuel product of claim 17,wherein the sintering the non-stoichiometric mixture and the fragilefuel particles includes: decomposing the hydride of the ZrH atapproximately 900 degrees Celsius during the sintering to achieve aprocessing pressure of less than or equal to 20 megapascals (MPa) andthe suppressed processing temperature between 1,650-1,800 degreesCelsius.
 19. The method of fabricating the nuclear fuel product of claim16, wherein the non-stoichiometric mixture includes the ZrH up to 10weight percent by mass and the carbon between 0.1 to 4 weight percent bymass.